The contribution of UPTF experiments to resolve some scale-up uncertainties in countercurrent two phase flow

The volume scaling of previous existing integral test facilities is up to a maximum of of the full size power reactors (or for the reflood integral test facility CCTF). The experimental results achieved in these facilities or in down-scaled separate effects test facilities have to be extrapolated to reactor scale in order to evaluate the full size reactor thermal hydraulic behaviour. There is some uncertainty in extrapolating the small scale results.Experimental results from the 1:1 scale UPTF test facility can considerably reduce the uncertainty of geometrical scaling. The final goal is to qualify the overall validity of computer program simulation, and to quantify the uncertainty of the computer program calculations for full size reactors.Significant scaling dependent experimental results have been found in UPTF separate effects tests in the case of heterogeneous steam-water flow conditions. This is especially the case for vertical and horizontal countercurrent flows.     

Performance and Upgrade of the DAFNE Beam Test Facility (BTF)

The DAFNE Beam Test Facility (BTF) at the Frascati National Laboratory of INFN (LNF) has been operational since October 2002. This facility provides electron and positron beams with tunable energy from 25 MeV to 750 MeV, while the intensity can be varied from$10^10/ pulse$@ 0–50 Hz down to a single particle. In the last two years many experiments and beam tests have been hosted, so that the large use of the facility pushed the DAFNE staff to upgrade both the beam line, in order to improve the duty cycle, limited by frequent beam injection in the DAFNE main rings, and the diagnostic devices. The beam characteristics, the user experience and their typical measurements, the diagnostic tools developed and the available equipment are presented. Finally, we report on the facility upgrade and on operation in the new configuration.     

CICC Joint Development and Test for the Test Facility

The superconducting joint of the NbTi Cable-in -conduit Conductor (CICC) has been developed and tested on the magnet test facility at Institute of Plasma Physics, Chinese Academy of Sciences. The CICC is composed of (2NbTi+lCu)x3x3x(6+ltube) strands each with 0.85 mm in diameter, which has been developed for a central solenoid model coil. The effective length of the joint is about 500 mm. There have been two common fabrication modes, one of them is to integrate the 2 CICC terminals with the copper substrate via lead-soldering, and the other is to mechanically compress the above two parts into an integrated unit. In the current range from 2 kA to 10 kA the joint resistance changes slightly. Up to now, 11 TF magnets, a central solenoid model coil, a central solenoid prototype coil, and a large PF model coil of PF large coil have been completed via the latter joint in the test facility.     

Experimental investigation and CFD validation of countercurrent flow limitation (CCFL) in a large-diameter PWR hot-leg geometry

ABSTRACT

Countercurrent flow limitation (CCFL) is a phenomenon that consists of several flow patterns occurring simultaneously which produces a complex gas/liquid interface and interfacial momentum transfer, thus making it one of the most challenging two-phase flow configurations for computational fluid dynamics (CFD) validation. Numerous experimental investigations have been carried out in recent years regarding this, but most of those investigations were performed in small-diameter pipes or in non-pipe geometries (rectangular cross sections). A review of these experimental investigations has shown that the scale and geometry of the test section has a large impact upon the onset and characteristics of the CCFL. In order to provide a better understanding of this phenomenon in an actual pressurized water reactor (PWR) hot-leg geometry at a relatively large-diameter and scale, a test facility with a ～1/3.9 scale and a 190 mm inner diameter was constructed. Experiments were carried out at atmospheric pressure using water and air. High-speed recording was used to acquire high-quality images of the air/water interface. CCFL mechanisms, flow patterns, and the limits of the onset of CCFL and deflooding were experimentally identified. CFD simulations of two representative cases were carried out and assessed against experimental results. The analysis of the CFD simulations has provided insights into the improvements necessary for the accurate simulation of CCFL in large-scale geometries.     

Piping Mechanics Analysis of Sodium Related Equipment Comprehensive Test Facility(F202)

正Sodium related equipment comprehensive test facility used to validated the mechanical sodium pump,sodium-water steam generator,control rod drive mechanism,elevator,sodium valve,sodium level meter calibration device,sodium flowmeter calibration device and other key sodium related equipment.The material of pipes is austenitic steel     

AC-3-irradiation test of sphere-pac and pellet (U,Pu)C fuel in the US Fast Flux Test Facility

The objective of the AC-3 bundle experiment in the Fast Flux Test Facility (FFTF) was to evaluate a fuel fabrication method by ‘direct conversion’ of nitrate solutions into spherical uranium–plutonium carbide particles and to compare the irradiation performance of ‘sphere-pac’ fuel pins prepared at Paul Scherrer Institute (PSI) with standard pellet fuel pins fabricated at Los Alamos National Laboratory (LANL). The irradiation and post test examination results show that mixed carbide pellet fuel produced by powder methods and sphere-pac particle fuel developed by internal gelation techniques are both valuable advanced fuel candidates for liquid metal reactors. The PSI fabrication process with direct conversion of actinide nitrate solutions into various sizes of fuel spheres by internal gelation and direct filling of spheres into cladding tubes is seen as more easily transferable to remote operation, showing a significant reduction of process steps. The process is also adaptable for the fabrication of carbonitrides and nitrides (still based on a uranium matrix), as well as for actinides diluted in a (uranium-free) yttrium stabilized zirconium oxide matrix. The AC-3 fuel bundle was irradiated in the Fast Flux Test Facility (FFTF) during the years 1986–1988 for 630 full power days to a peak burn up of 8 at.% fissile material. All of the pins, irradiated at linear powers of up to 84 kW/m, with cladding outer temperatures of 465 °C appeared to be in good condition when removed from the assembly. The rebirth of interest for fast reactor systems motivated the earlier teams to report about the excellent, still perfectly relevant results reached; this paper focusing on the sphere-pac fuel behaviour.     

核反应堆试验台架比例分析方法的发展和应用

比例分析方法为建立合理的反应堆安全系统缩比试验台架提供了理论基础。本文结合比例分析方法的发展,探讨了不同比例方法的特点,并总结了部分已有台架的比例设计概念及评价,为反应堆系统试验台架比例方法的选取提供了参考。结果表明,线性比例方法中的加速度比例项使其应用受到限制;功率-体积法是一种简单有效的比例方法,但瘦高台架的特点也使此方法存在不可避免的弱点;H2TS（HierarchicalTwo-TieredScaling）方法以PIRT（PhenomenaIdentificationRankingTable）表为基础,对系统中重要整体过程和局部过程均进行了比例分析,其发展的相似准则中含有流体物性比例项,为台架比例概念的发展提供了条件。我国将以H2TS方法为指导建立非能动堆芯冷却系统试验台架ACME。     

非能动试验台架中CMT的比例分析及失真评价

堆芯补水箱（CMT）是AP1000非能动堆芯冷却系统中的关键设备,对其进行合理的比例分析对非能动整体性能试验台架的设计起着重要作用。采用H2TS比例分析方法对CMT的循环模式和排水模式进行比例分析,进而将得到的CMT重要过程的相似准则应用于我国正在设计建造的ACME台架的CMT比例设计,并对其特征 Π 群的比例失真度进行定量化计算。最后,对ACME台架的CMT进行比例失真原因分析和评价。结果表明,CMT循环阶段的主要过程能在ACME中得到较好的模拟,而在排水阶段由于ACME超比例的CMT金属质量引起的储冷问题导致蒸汽冷凝过程存在一定的失真,但综合分析认为ACME台架采用高压模拟方案能较好地复现原型电站CMT的重要现象和过程。     

极端条件下辐射与物质相互作用(MaRIE)研究装置概况

极端条件下的辐射与物质相互作用实验装置,是洛斯·阿拉莫斯国家实验室针对美国核武器库存维护计划需要,正在建设的旗舰级实验装置,其主要任务是促进对核武器相关材料的鉴定、认证和评估。本文从建设目的、组成部分和发展规划等方面简述了该装置的概况,并结合美国核库存维护目前面临的挑战,分析了该装置的建设对美国核库存维护的重要意义。     

Initial Thermal-Hydraulic Behaviors under Simultaneous ECC Water Injection into Cold Leg and Upper Plenum in a PWR-LOCA

Core thermo-hydrodynamic characteristics under the combined injection mode before and just after the beginning of bottom reflood of a PWR-LOCA were experimentally studied by performing three tests in Slab Core Test Facility simulating a full radius slab section of a PWR. Emergency core cooling water was simultaneously injected into the upper plenum and the intact cold leg. The subcooling and the radial distribution of the upper plenum injection water were the test parameters.

The core was cooled by falling water before the beginning of bottom reflood. However, the core was finally quenched by bottom reflood. Before the beginning of bottom reflood, the transients of water level in the lower plenum were different among three cases, that is, the water level was rapidly or gradually increased in the first and second cases, respectively, or remained below the bottom of core barrel in the third case. The bottom reflood was much delayed in the last two cases. Even under the conditions with large upper plenum injection rate of subcooled water and with steam escape through the lower plenum, continuous fall back was not observed but the subcooled water was intermittently supported by the upward steam flow generated in the core.     

Upgrading the data acquisition and control systems of the European Breeding Blanket Test Facility

The EBBTF (European Breeding Blanket Test Facility) experimental plant is a key component for the development of the breeding blankets (TBMs test blanket modules, HCLL helium cooled lithium lead and HCPB helium cooled pebble bed types) used by ITER. EBBTF is an experimental plant which provides the double breeding/cooling loops (liquid metal and gas) required for HCLL testing. EBBTF is composed of four subsystems (TBM, IELLLO integrated European lead lithium loop, HE-FUS3 helium fusion loop, version 3 and helium compressor build by ATEKO) with dedicated control systems realized with hardware/software combinations covering 15 years (1995–2010) time span. At the end of 2010 we began to upgrade the HE-FUS3 data acquisition control systems (DACS) replacing the obsolete PLC Siemens S5 with National Instruments Compact FieldPoint and LabVIEW. The control room has been completely reorganized using high resolution monitors and workstations linked with standard Ethernet interfaces. The data acquisition, control, safety and SCADA software has been completely developed in ENEA using LabVIEW. In this paper we are going to discuss the technical difficulties and the solutions that we have used to accomplish the upgrade.     

Experimental investigation on similarity between velocity and density profiles in density-stratified countercurrent flow in reactor horizontal leg

A density-stratified countercurrent flow was investigated to obtain data necessary to develop a physical model on a thermally stratified flow in a horizontal leg of a pressurized water reactor (PWR). The experiments were conducted at atmospheric pressure and temperature using fresh water and NaCl solution with a non-dimensional density ratio of up to 1.2. The emphasis was placed on measurements of velocity and concentration profiles near the interface between the two fluid layers. Measured mean velocity and concentration profiles were fitted consistently using the Monin–Obukhov similarity theory, which are well-known outcomes for stratified turbulent shear flow. The interfacial friction and entrainment coefficients obtained from the fitted profiles agreed well with existing results in literature, confirming the applicability of the Monin–Obukhov theory. Furthermore, a new empirical correlation was proposed for the prediction of a mixing layer thickness.     

Validation of a Computational Simulation Method for Evaluating Thermal Stratification in the Reactor Vessel Upper Plenum of Fast Reactors

Validation of a numerical simulation method is carried out for thermal stratification phenomena in the reactor vessel upper plenum of advanced sodium-cooled fast reactors. The study mainly focuses on the fundamental applicability of commercial computational fluid dynamics (CFD) codes as well as an inhouse code to the evaluation of thermal stratification behavior including the simulation methods such as spatial mesh distribution and RANS-type turbulence models in the analyses. Two kinds of thermal stratification tests are used in the validation, which is done for relatively simple- and conventional-type upper plenum geometries with water and sodium as working fluids. Quantitative comparison between the simulation and test results clarifies that when used with a high-order discretization scheme of the convection term, the investigated CFD codes are applicable to evaluations of the basic behaviors of thermal stratification and especially the vertical temperature gradient of the stratification interface, which is important from the viewpoint of structural integrity. No remarkable difference is seen in the simulation results obtained using different RANS turbulence models, namely, the standard kε model, the RNG k-ε model, and the Reynolds stress model. It is further confirmed in a numerical experiment that the distribution of two or more meshes within the stratification interface will lead to accurate simulation of the interface temperature gradient with less than 10% error.     

Preliminary design of the control and data acquisition systems for the Neutral Beam Test Facility

The Neutral Beam Test Facility, which will be built in Padova, Italy, is aimed at developing the ITER heating neutral beam injector (HNB) and at testing and optimizing its operation up to nominal performance before installation on ITER. It requires the development of two independent experiments referred to as SPIDER (source for production of ions of deuterium extracted from Rf plasma) and MITICA (megavolt ITer injector & concept advancement). SPIDER will explore the full-size negative ion source for ITER, whereas MITICA will explore the full-size ITER neutral beam injector. Both experiments will be designed for long-pulse operation, up to 3600 s, as ITER itself. MITICA includes three functional components: the heating neutral beam injector plant system (HNB), which is the device under test; the auxiliary plant system (AUX), which includes all equipment to operate the HNB in the test facility (e.g. the local electric grid to feed the HNB power supplies), and MITICA supervisory system that is an electronics/informatics infrastructure to operate the facility. The paper introduces the requirements for the control and data acquisition systems of the experiments and proposes a preliminary design for both systems. SPIDER, which is preparatory to MITICA and will be developed on a shorter time scale, has no constraints coming from ITER CODAC, whereas MITICA includes the ITER neutral beam injector and therefore must be fully compatible with ITER CODAC.     

The influence of walls and upper tie plate slots on the flooding mechanism in fuel elements with and without heat transfer between steam and water

The counter-current flow of steam and water was experimentally investigated for the upper part of a PWR fuel element. The actual geometrical shape of the nuclear equipment was simulated by various types of plates, in which bore holes and slots were arranged in different positions. The experiments were performed with and without an installed, unheated rod bundle below the plates. The water was injected at saturated and subcooled temperatures in order to observe the effects of heat transfer on counter-current flow.

With increasing steam velocity the flooding occurs initially in the tie-plate area. If the rod bundle is installed in the flow duct, a part of the downwards flowing water is transported upwards from the region of the upper grid spacer to the plate. Heat transfer between the phases can cause in the counter-current flow region an instable transition from downward to near complete upward directed liquid flow. In comparison to experiments with saturated water injection, flooding occurs at larger steam velocities. Different flooding correlations, which are known from the literature, were compared with the experimental data to appraise their applicability to counter-current flow in the core of PWRs.     

兰州重离子研究装置辐射环境安全性评估

描述了兰州重离子研究装置(HIRFL)运行10多年来对其周围中子、γ辐射水平，环境中水、土、植物等介质的总α、总β放射性水平以及屏蔽周围的中子、γ辐射水平和工作人员个人剂量的测量结果，并进行了初步评价。监测结果表明，兰州重离子研究装置的运行没有造成环境的放射性污染，运行是安全的。     

Tests of a Two-Stage Core at the KV-1 Test Facility: Physical Aspects and Results

Atomic Energy - The physical aspects and main results of reactor tests of a two-stage core consisting of fresh fuel assemblies and a significant number of fuel assemblies from the previous core,...     

Primary Investigation into the Laser Triggering Multi-Gap Multi-Channel Gas Switch in a Single Test Module Facility

A high precision laser trigger system is built up in the single test module of Primary Test Stand （PTS） facility. A fourth harmonic, with a wavelength A of 266 nm, Q-switched Nd：YAG laser was used to trigger the 5 MV multi-gap multi-channel gas switch which was filled with high pressure SF6-N2 mixture gas. The maximum deviation and the standard deviation in the jitter time of the trigger system is 4- 0.7 ns and 0.3 ns respectively. The maximum deviation and the standard deviation in the jitter time for the multi-gap multi-channel laser triggering switch is 4- 2.4 ns and 1.5 ns respectively. The curve of switch delay-time versus laser energy is obtained, which is helpful for the choice of fitting laser energy. The successful test with two lasers indicated that the design of using twenty-four lasers to trigger twenty-four switches respectively is feasible in ＂PTS＂.     

高温气冷实验堆堆芯出口热气联箱混合性能对比试验

１０ＭＷ高温气冷实验堆堆芯出口冷却剂温度径向分布很不均匀，若不使之均匀化，将造成蒸汽发生器件部上过大的热应力，设置在堆底反射层中的堆芯出口热气联箱的作用之一是使冷却剂氦气在其中得到充分的热混合。     

Radiochemical Analysis Auto-system of ~(90)Sr in Liquid Effluent from Nuclear Facility

<正>The analysis of~(90) Sr is an important content in liquid effluent from nuclear facilities and environmental radioactivity monitoring.The traditional analytical methods include fuming nitric acid method,ion exchange,solid-phase extraction method,etc.These analysis methods all relyon manual analysis.In order to further reduce the labor intensity and workload of analysts,an automatic analysis system of 90 Sr in environmental water was developed based on the solid phase extraction technology.